Reactor vessels in Pressurized Water Reactors (PWRs) are required by law to be evaluated for a Pressurized Thermal Shock (PTS) event. The PTS event consists of a rapid cooldown and depressurization of the reactor vessel and a subsequent repressurization. The concern to be addressed is that integral steel (typically alloys are A501, A508 and A533) reactor vessels become embrittled in the beltline region due to neutron bombardment. See A Guide to Nuclear Power Technology,.COPYRGT. 1984 John Wiley and Sons, Inc., New York, N.Y. (Chapter 10, pgs. 403-449) Reprint Edition 1992 by Kreiger Publ. Co., Malabar, Fla. 32950). As the vessel wall loses its ductility, a PTS event may initiate a crack or extend a pre-existing crack in the vessel wall. Some vessel materials, including weld joints, consist of certain chemical compositions that make them more susceptible to irradiation damage (embittlement) than others. For these vessels, current evaluation methodologies may not demonstrate sufficient crack resistant behavior during a PTS event to satisfy regulatory requirements. If this occurs, the industry currently does not have an accepted, proven methodology to demonstrate vessel integrity. The Yankee Rowe nuclear plant shut down in 1992 due to these concerns. The suggested fixes for the Yankee Rowe steel vessel were deemed to be both too risky and too expensive to adopt.
All plants currently have a flux reduction program in place to reduce the rate of neutron embrittlement in the vessel. These programs include fuel management and local shielding efforts. The effect of these programs is to extend the operating life of the reactor vessel by postponing the time when the level of potential embrittlement reaches a regulatory screening level for PTS. Even with these flux reduction programs, however, some irradiated vessels will exceed the PTS screening criteria prior to the plant's operating life goals. As with the Yankee Rowe plant, this may result in a premature shutdown of an operating nuclear plant.
Once the PTS screening criteria is exceeded, there are two courses of action previously considered in the industry. The first is a complex probabilistic fracture mechanics and event three analysis whose goal is to demonstrate that the probability of unacceptable cracking in the vessel wall is below acceptance criteria. This analytical approach is not currently considered credible because of its complexity, inherent assumptions, and dependence on input data that is not readily or reliably available. Much of these uncertainties revolve about the actual material condition and properties at critical locations in the reactor vessel wall.
The second approach which has been considered is to anneal the irradiated reactor vessel in-situ to regain the material ductility lost to irradiation embrittlement. Annealing has not been performed on a commercial nuclear plant reactor vessel in the U.S.A. despite the teachings of U.S. Pat. No. 4,652,423, which are included herein by reference. Annealing has been performed in the Soviet Union, but those vessels are smaller, of different construction, and of different materials. Annealing is considered a high risk approach because it could result in permanent deformation or damage to a vessel which could end its useful life. The rate of re-embrittlement is also in question. In addition, the regulatory risk is high. The cost of an in-situ anneal has been estimated to be in the fifteen million United States Dollars ($15M USD) to twenty-five million United States Dollars ($25M USD) range and may be cost prohibitive for many plants even if the technical and regulatory risks are overcome.
The concept of mechanical prestress has been applied to other situations in the past. Cable wraps about cannon barrels were used on British warships to prevent blowing apart cannons when the forged or cast barrels were of poor quality. In Boiling Water Reactors (BWRs), stress corrosion cracking of stainless steel piping has been retarded by utilizing a weld overlay technique or clamping band that applies a compressive load to the piping as the weld overlay cools and shrinks. As explained in U.S. Pat. No. 3,742,985, these prior art methods can be likened to the same thing as occurs to a wooden wheel when a blacksmith "sweats-on" a steel or iron rim and the wheel becomes tightened as shrinkage of the cooling iron or steel occurs.